1. Field of the Invention
This invention relates to (1) a process for volume reduction and solidification of a radioactive waste solution mainly composed of sodium borate and discharged from a pressurized water reactor (hereinunder abbreviated as "PWR"), which comprises adding, to said radioactive waste solution or its concentrate or the dried material of the concentrate, ZnO or a mixture of ZnO with Al.sub.2 O.sub.3 and/or CaO, and then subjecting the resulting mixture to dehydration and melting by heating to produce a vitric solid solution composed mainly of Na.sub.2 O-B.sub.2 O.sub.3 -ZnO, (2) a vitric solid solution produced by said process, and (3) an apparatus for producing said vitric solid solution.
2. Description of Prior Art
As processes for solidification of sodium borate waste solutions discharged from PWRs, there have hitherto been used the cement solidification process and the bituminization process.
In the cement solidification process, water in waste solution remains in solidified wastes and accordingly these solidified wastes have volumes larger than the original waste solutions. In the bituminization process, firstly water is removed by evaporation from waste waters and then molten bitumen is added to the evaporation residues to solidify the residues. Therefore, the bituminization process is fairly effective in volume reduction and has been employed at many PWR plants.
Solidification of radioactive wastes has been conducted on the assumption that solidified wastes can in some cases be dumped into the ocean. However, ocean dumping of solidified wastes is becoming increasingly prohibitive and also their inland disposal or underground burial is difficult to realize in countries without vast disposal sites. For these reasons, solidified radioactive wastes are compelled to be stored within the site of nuclear power stations over a very long period of time. Since each site has a limitation in available space, a solidification process which is more effective in volume reduction than the conventional solidification processes is required.
Presently, in the bituminization process for PWR sodium borate waste solutions, a solidified waste is produced so that the solid content, apart from asphalt, in the solidified waste becomes about 40% by weight. If the content of solid material other than asphalt could be increased to 60% by weight, the effect of volume reduction becomes larger. This comparison is illustrated more intelligibly in FIGS. 1 and 2, wherein the solidified waste of 40% solid content and that of 60% solid content are compared on the basis of volume. FIGS. 1 and 2 each illustrate a flow diagram of solidified waste production by using the volume of the total solid content of a concentrated waste solution supplied and the volume of asphalt. FIG. 1 shows the case of production of a solidified waste of 40% solid content and FIG. 2 shows the case of production of a solidified waste of 60% solid content. In each case, the concentrated waste solution supplied had a boron concentration of 20,000 to 21,000 ppm and a total solid concentration of about 10%, and the sodium borate in the concentrated waste solution had a density of about 2.36.
Even if the solid content, without asphalt, in the above solidified waste could be increased to 60% by weight, problems would still exist. That is, the volume of asphalt as a binder relative to the solid content of the concentrated waste solution composed mainly of sodium borate crystal occupies more than 50% of the volume of the solidified waste, and, further, the extent of swelling of the solidified waste in water can not be ignored and addition of anti-swelling agents such as CaO or BaO becomes necessary.
The plastic solidification process is also proposed. In this process, the solid content without plastic in solidified waste is reported to be roughly 60% by weight. However, a large amount of an additive such as Ca is said to be required at the time of solidification of sodium borate waste solution.
Asphalt has a relatively low flash point. Plastics are organic substances though some of them have high flash points. Hence, neither asphalt nor plastics are sufficiently free from the risk of fire occurring during their transportation in the form of solidified waste as a result of vehicle collision, etc. From the standpoint of fire safety, solidification of radioactive waste solutions by the use of an inorganic substance is desirable.
Based on the thinking that the effect of volume reduction is enhanced if asphalt or a plastic as a binder is removed, a process was developed for volume reduction of sodium borate water solution by roasting or calcinating a sodium borate waste solution on a fluid bed of a temperature of several hundreds degrees centigrade, as described in Japanese Patent Laid-open Publication No. 158998/1981 and Japanese Patent Laid-open Publication No. 30000/1982. However, calcinated sodium borate must be protected from moisture or water penetration during storage, and therefore it can not be disposed of as it is.
It is known that sodium borate, regardless of whether it is radioactive or not, is very easily converted to a vitric substance when it is melted by itself. By varying a Na.sub.2 O to B.sub.2 O.sub.3 ratio, vitric substances of various compositions are formed and all these substances are readily soluble in water.
Besides this, as a solidification process using boric acid as a solidifying material, there is a solidification-in-borosilicate-glass process which is intended to be used as a solidification process for highly radioactive waste solutions from nuclear fuel reprocessing plants. (New Technilogies Viewed from Patents--16, in 1983 May issue of "Invention").
The above borosilicate glass process can provide safe solidified wastes but it is not desirable in terms of volume reduction of PWR sodium borate waste solution.
In Table 1 are shown examples of compositions of borosilicate glasses generally known in Japan. The table was summarized based on the publications listed on the page following Table 1.
TABLE 1 __________________________________________________________________________ Compositions of Solidified Borosilicate Glass Wastes Experimentally Produced by Power Reactor and Nuclear Fuel Development Corporation and Japan Atomic Energy Research Institute Reprocessing at Power Reactor and Nuclear Fuel Development Japan Corporation Atomic Glass Basic composition Energy component for engineering tests Glass composition: G-2 GB-1 Research % by wt. Range Standard (1) (2) (3) (4) (1) (2) GB-4 Institute __________________________________________________________________________ SiO.sub.2 35 to 45 43 43 43 43.9 43.0 53.4 54.7 50.5 27.0 B.sub.2 O.sub.3 10 to 18 14 14 14 14.3 9.4 9.4 9.6 13.4 30.3 Al.sub.2 O.sub.3 to 5 4 4 4 4.1 3.5 4.9 5.0 5.3 -- Na.sub.2 O 9 to 12 10 10 10 9.98 10.18 10.18 10.18 11.18 12.5 K.sub.2 O to 4 2 2 2 2.0 2.0 -- -- -- Li.sub.2 O 2 to 5 3 3 3 3.1 3.0 1.9 1.0 1.0 -- CaO to 3 2 2 2 2.0 2.0 -- -- -- 1.7 ZnO to 8 2 2 2 2.0 2.0 -- -- -- -- Fe.sub.2 O.sub.3 -- -- -- -- -- 0.97 0.97 -- -- -- Cr.sub.2 O.sub.3 -- -- -- -- -- 0.17 0.17 -- -- -- NiO -- -- -- -- -- 0.12 0.12 -- -- -- F.P. oxide 20 20 20 20 18.7 19.1 19.1 19.5 18.6 28.5 Publication (1) (2) (3) (4), (5) (6) (6) (4), (5) (7) __________________________________________________________________________ Publication: (1) Summary Report F 25 (March, 1978), of 1978 Annual Meeting of Atomic Energy Society of Japan (2) Preprint G 46 (October, 1978), of 1978 Fall Meeting of Atomic Energy Society of Japan (3) Summary Report J 37 (March, 1979), of 1979 Annual Meeting (17th) of Atomic Energy Society of Japan (4) Summary Report J 39 (March, 1979), of 1979 Annual Meeting (17th) of Atomic Energy Society of Japan (5) Preprint F 74 (October, 1979), of 1979 Fall Meeting of Atomic Energy Society of Japan (6) Preprint P 58 (October, 1980), of 13th Reporting and Lecture Meeting of Power Reactor and Nuclear Fuel Development Corporation (7) Perprint G 47 (October, 1978), of 1978 Fall Meeting of Atomic Energy Society of Japan Note: According to the Power Reactor and Nuclear Fuel Development Corporation, leaching rates, etc. of various borosilicate glass wastes were evaluated by 1978 and, as a result, G2 was selected as a satisfactory composition; the effects of CaO and ZnO on leaching rate were not conspicuous but BaO and TiO.sub.2 were very effective with addition of only a small quantity. (Summary Report F 31 of 1981 Annual Meeting of Atomic Energy Society of Japan)
The glass composition G-2 in Table 1 is regarded as a typical composition of borosilicate glass solidified wastes to be produced at Japanese nuclear fuel-reprocessing plants. As is obvious from Table 1, in G-2, SiO.sub.2 is roughly about 50% by weight, B.sub.2 O.sub.3 is about 14% by weight and Na.sub.2 O is as low as about 10% by weight. Hence, if a vitric solidified waste of the same composition as that of G-2 is actually produced from a PWR sodium borate waste solution, the vitric solidified waste is very poor in volume reduction effect.
FIG. 3 is a three component systems diagram showing compositions of borosilicate glass frits, of presently known waste solutions from reprocessing of nuclear fuels. In these frits, including those being used in U.S.A. and other foreign countries, the respective contents of B.sub.2 O.sub.3 and Na.sub.2 O are very low while the content of SiO.sub.2 is high. For comparison, there is also shown in FIG. 3 a composition of a Na.sub.2 O-B.sub.2 O.sub.3 -ZnO type (free from SiO.sub.2) vitric solidified waste frit of the present invention. Incidentally, FIG. 3 has been taken from "Preprint F 9" of the Fall Meeting, 1980 of the Atomic Energy Society of Japan.
In the section on borosilicate glass of "3.2 Vitrification of Radioactive Wastes" of the Glass Handbook, there is shown a borosilicate glass produced in West Germany which contains boric acid and sodium in fairly high quantities, namely, 17 to 20 mole % of B.sub.2 O.sub.3 and 10 to 25 mole % of Na.sub.2 O. However, containing a large quantity of SiO.sub.2, the borosilicate glass is not fully satisfactory in volume reduction effect.